ML20323A037

IMC 2515 Appendix D Plant Status

ML20323A037
Person / Time
Issue date:	01/13/2021
From:	Aron Lewin NRC/NRR/DRO/IRIB
To:
Aron Lewin, NRR/DRO/IRIB, 415-2259
Shared Package
ML20325A102, ML21014A003	List:
References
CN 21-004, DC 20-042
Download: ML20323A037 (19)

Text

Issue Date: 01/13/21 1

2515 Appendix D NRC INSPECTION MANUAL IRIB INSPECTION MANUAL CHAPTER 2515 APPENDIX D PLANT STATUS Effective Date: 01/13/2021 2515D-01 PURPOSE The Reactor Oversight Process recognizes that resident inspectors have a specific responsibility, outside of inspection activities, to be aware of plant conditions on a routine basis.

This appendix provides guidance regarding these plant status activities at pressurized water reactors (PWRs) and boiling water reactors (BWRs).

Resident inspectors knowledge of plant activities and status is important in the risk-informed inspection process for determining how to select and implement the appropriate baseline inspection procedures. Plant status activities should focus on being aware of emergent plant issues, potential adverse trends, current equipment problems, and ongoing activities, including their impact on plant risk. Based on the knowledge gained through the plant status review, the inspectors are expected to make adjustments to their inspections so that they can inspect activities which are of higher risk significance. Included in these activities is the awareness of how licensees are managing fatigue due to the impact this can have on the protection of public health and safety and common defense and security. Additionally, resident inspectors should periodically (once a quarter) conduct tours of security-related areas in order to identify any security-related issues which may warrant follow-up by region-based security inspectors.

The resident inspector should transition into the appropriate inspection procedure whenever their effort shifts from collecting status information to evaluating a potential inspection issue.

Security-related issues identified during tours of the licensee facility should be referred to security specialists in the region for follow-up inspection(s) as appropriate. The inspector should transition into the appropriate inspection procedure if the information collection activity will exceed about 1/2 hour for any single issue. Scope of activities conducted under the Plant Status procedure does not require documentation in inspection reports.

The frequency of the plant status review effort should be determined by the inspector based on current plant conditions and activities. Inspectors should use plant-specific risk information to determine what systems and activities are of higher risk significance given the present plant configuration.

2515D-02 OBJECTIVES 02.01 To be aware of plant conditions on a routine basis.

Issue Date: 01/13/21 2

2515 Appendix D 2515D-03 APPLICABILITY See Section 2515-03 of Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase.

2515D-04 DEFINITIONS None.

2515D-05 RESPONSIBILITIES AND AUTHORITIES See Section 2515-05 of IMC 2515, Light-Water Reactor Inspection Program - Operations Phase.

2515D-06 REQUIREMENTS See Section 2515-11 of IMC 2515, Light-Water Reactor Inspection Program - Operations Phase.

2515D-07 GUIDANCE 07.01 Control Room Walkdown The purpose of the control room walkdown is to help enable the inspector to stay current of plant status as well as to identify unexpected plant conditions that warrant additional inspection under the baseline inspection program. Evaluate the status of the safety or risk important systems by observing the indicated parameters and equipment configuration indications on the control boards. This walkdown is intended to be general (not detailed) in nature.

See Inspection Procedure (IP) 71153, Event Follow-up, Appendix B. It provides guidance on NRC inspector conduct while in the control room during events in order to preclude NRC intrusion in licensee response activities.

Examples of items for which inspectors should identify during control room walkdowns include:

1. Look for system components that are in unexpected configurations or parameters that are at unexpected values based on the operational mode of the plant

2. Identify any alarming or locked-in annunciator conditions.

3. Note whether any adverse plant parameter trends exist and whether the licensee is aware of the trends.

4. Identify whether the plant is in any technical specification (TS) limiting conditions for operation (LCOs), whether the TS action statements are being met, and those TS requirements and license conditions are being met.

Issue Date: 01/13/21 3

2515 Appendix D

5. Determine if the licensee is operating with multiple, repetitive, or unplanned TS action statement entries caused by degraded equipment conditions; that they are assessing and managing the risk associated with this condition in accordance with licensees procedures; and that the issue associated with the degraded equipment conditions is entered into the corrective action process.

6. Determine if the licensee is operating within licensed power levels. Guidance for evaluating brief power level fluctuations above 100% is given in NRC Regulatory Issue Summary (RIS) 2007-21, Adherence to Licensed Power Limits, Revision 1 (ML090220365).

7. Any radiation dose implications associated with repetitive tasks should be reviewed by applicable radiation safety baseline IPs. In the control room or other appropriate locations, review visible portions of radiation monitors or other indications that could provide indication of an apparent uncontrolled release.

8. Observe equipment and / or operator actions for Open Phase Condition monitoring as specified by licensee commitments or NRC requirements Review control room logs, equipment out-of-service or clearance logs, compensatory action logs, TS logs, chemistry logs, standing orders, and night orders several times each week to become aware of potential risk-related problems that occurred since the previous review.

Licensees may refer to compensatory actions as Operator Work Arounds, Operator Burdens, etc. Determine whether the logs appropriately reflect the plant status observed during the control board walkdown and whether TS requirements are being met. A review of the operator shift logs and standing orders may provide insights regarding equipment operability. Pursue any operability or functionality concerns using IP 71111.15, Operability Determinations and Functionality Assessments.

Report primary-to-secondary leakage in steam generators which is greater than three gallons per day to NRC headquarters staff. For additional information on the reporting guidance, see Section 07.05 of IMC 0327, Steam Generator Tube Primary-To-Secondary Leakage.

If the licensee documents waivers of work-hour controls in the control room logs or shift manager logs, then periodically review the waiver(s) to determine that the granting of the waiver(s) addressed circumstances that could not have been reasonably controlled. If further inspection guidance is needed then IP 93002, Managing Fatigue, may be referenced on an as needed basis.

It is important that inspectors maintain awareness of situations that may result in increased fatigue (i.e., unit outages, short duration LCOs, staff shortages, etc.). When evidence of fatigue is identified, inspectors should immediately notify licensee management of any observed condition that indicates signs of fatigue so they can evaluate the need for a fatigue assessment per 10 CFR 26.211, Fatigue Assessments.

To ensure that the licensee properly monitors for Reactor Coolant System (RCS) pressure boundary leakage or potential unidentified leakage exceeding TS limit, the inspector should routinely determine if the licensee:

Issue Date: 01/13/21 4

2515 Appendix D

1.

Monitors leak detection systems such as the containment atmosphere particulate radioactivity instruments, the containment sump flow/level instruments, the containment atmosphere gaseous radioactivity instruments, the containment humidity instruments, and/or any plant-specific instrumentation to indicate potential RCS leakage. Also, trends these parameters for potential adverse trends.

2.

Takes appropriate actions for degraded or inoperable leak detection instrumentation or alarms in accordance with TS and responds to alarms in accordance with alarm response procedures. Also, periodically verifies that the alarm response procedure actions are consistent with plant licensing documents.

3.

Periodically performs the inventory balance check (PWR only) and attempts to confirm RCS unidentified leakage with alternate and diverse means, such as, changes in containment sump level or sump pumping frequency and volume.

4.

Takes appropriate actions in accordance with plant-specific leak rate impact or leakage investigation procedures (leakage source identification, quantification, classification, etc.) when RCS leakages are suspected. Also, considers unidentified leakage as identified leakage only when the leak rate has been actually measured and identified.

5.

Conducts activities to identify sources of RCS unidentified leakage. Documents actions taken to identify sources of unidentified RCS leakage in the control room logs or in the corrective action program, as specified in plant administrative procedures. The licensees leak identification plan includes actions such as system walkdowns; system surveillance and re-alignment; containment entry (PWR only) and visual inspections for boric acid deposits (PWR only); verification of pumps and valves for possible seal and packing leakages; inspection of pipe flanges and major welds, including instrument lines and connections; and sampling/ performing isotopic analysis of atmospheres, filter elements, and sumps.

6.

Trends unidentified leak rates, pays particular attention to changes in unidentified leakages, and takes appropriate corrective action for adverse trends. Also, trends other containment parameters such as containment sump inleakage rates, the containment air/gaseous radiation monitor indication, the containment particulate radiation monitor indication, and the containment humidity indication to validate potential RCS unidentified or pressure boundary leakages.

If the inspector observes significant adverse trends, the inspector should engage licensee and regional management and the appropriate Office of Nuclear Reactor Regulation (NRR) technical branches as outlined in Attachment 1. As applicable, the inspectors should also determine if the licensee enters the appropriate procedure for responding to adverse RCS leakage trends.

Review licensee procedures for action steps, as unidentified leakage approaches licensee administrative limits or technical specifications allowed values. The inspector should use IP 71111.22, Surveillance Test, to verify licensees surveillance activities and IP 71111.04, Equipment Alignment, to conduct any plant walkdown. Review any operational and technical decision-making activities and pursue any operability concerns using IP 71111.15, Operability Evaluations. In addition, Attachment 1 provides a technique to aid inspectors in independently determining whether an adverse trend exists with licensees RCS unidentified leakage rate data obtained during steady state power operation.

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2515 Appendix D This guidance also provides action level criteria to assess the significance of the trend and licensees actions in response to increasing levels of unidentified RCS leakage that could indicate Reactor Coolant System pressure boundary (RCPB) degradation. This guidance is provided in response to Davis Besse Lessons Learned Task Force Report recommendation 3.2.1(2) (ML022760414).

07.02 Status Meetings Select and attend licensee meetings, on a routine basis, that provide an overall status of the plant and pertinent ongoing activities. These meetings could include the licensee's plan of the day meeting, shift turnover meeting, emergent work meeting, equipment prioritization meeting, and corrective action document review meeting. Note that during or in preparation phases of the plant refueling or maintenance outages, licensees may conduct additional meetings.

Inspectors should attend these meetings to understand the scope, schedule, and risk-significant activities of these outages. This will enable the inspectors to plan and implement applicable baseline inspection procedures that needed an outage. Additionally, the inspector should be aware that work-hour controls may change with a unit in an outage and an increase in the use of waivers, self-declarations, or fatigue assessments may occur.

The purpose of attending the status meetings is to gather information about overall site activities in order to determine what activities will be or are being conducted so that inspection resources can be appropriately focused on those activities with the higher safety significance.

07.03 Plant Tours On a weekly basis, tour accessible areas of the plant containing safety significant structures, systems, and components (SSCs), areas that contain significant radiological hazards, and areas with important physical security equipment. Focus on areas of the plant that inspectors have not entered while performing other inspections on a weekly basis.

Inspectors should coordinate with the licensee to tour areas which become accessible on an infrequent basis and for short periods of time to assess the material condition and status of safety systems, structures and components. While some areas not normally accessible might be obvious such as heater bays in BWRs, other areas may take additional effort to identify and plan for a tour (such as essential service water, radwaste vaults, outdoor underground vaults).

The inspectors should review and discuss areas not normally accessible with the licensee to ensure the inspectors are aware of their existence (some areas may not be obvious) and plan logistics such as ensuring advance notification of when they will be accessible, if appropriate, and any special arrangements needed for entry (i.e., special training for fall protection or confined space entry).

Inspectors should plan to tour areas not normally accessible and these inspections should coincide with the licensees schedule for accessing the area. Inspectors should place the highest priority on areas that contain risk significant or safety-related equipment but may take into account areas which contain equipment that could cause a transient or initiate a radioactive release. The inspectors can also review the results of licensees direct observations (video movies, and digital photographs) when direct inspections by inspectors were not possible or if other factors such as personnel safety or the radiation levels in the area to be inspected warrant use of licensees direct observations. It is not the intent of this guidance to force licensees to make every not normally accessible area of the plant accessible for NRC inspection.

Issue Date: 01/13/21 6

2515 Appendix D During changing plant conditions (plant refueling or maintenance outages), the frequency and scope of plant status tours may be increased to tour areas not normally accessible and to observe material condition and equipment in an abnormal lineup.

Plant tours should occasionally include off-site and on-site emergency response facilities, independent spent fuel storage facilities, and storage locations for equipment used for diverse and flexible mitigation strategies (FLEX). In addition, the inspector may accompany a plant operator performing equipment rounds to gain insights regarding undocumented plant deficiencies, work arounds, or temporary modifications.

The purpose of the tours is to provide an independent evaluation of ongoing plant activities that may affect plant performance in the cornerstones. In performing the tours, the inspector should keep in mind the integrated effect of plant problems on plant safety. Areas to note include:

1.

Plant activities taking place that may affect the operability of the required SSCs and/or increase plant risk including on-line (pre-outage) maintenance activities, such as the erection of temporary scaffolding, the installation of temporary services, and/or placement of other structures or material that may interfere with the safety-related function of SSC.

2.

The overall status of plant SSCs, including general material condition or the installation of unauthorized modifications that could affect the SSCs function. Pursue any unauthorized or temporary modification deficiencies using IP 71111.18, Plant Modifications.

A degraded condition is one in which the qualification of an SSC or its functional capability is reduced. Examples of degraded conditions are failures, malfunctions, deficiencies, deviations, and defective material and equipment. Examples of conditions that can reduce the capability of a system are aging, erosion, improper operation, and inadequate maintenance.

Obvious signs of degraded material condition of piping or other components, such as substantial corrosion, loose anchor bolts, leakage, standing water accumulation, cable insulation cracked or charred, or other conditions, may call into question operability or design margins of the equipment.

Inspectors should ensure that identified material condition deficiencies are captured in the licensees corrective action program. Inspectors should consult with appropriate regional and headquarters specialists if there are any questions regarding the operability or adequate design margin associated with degraded safety systems, structures, or components. Inspectors should attempt to obtain video movies and/or digital photographs of the degraded equipment (either on their own or through the licensee) to assist the specialists in evaluating the degraded material condition.

Inspectors should consider the potential for long-term degradation of SSCs or acceptance of long-standing degraded SSCs, as indicated by multiple similar entries in the licensees corrective action program. The licensees evaluation and resolution of such degraded SSCs should be considered for further inspection utilizing the appropriate baseline inspection procedure. For example, use-as-is determinations, revision of engineering or operational acceptance criteria, reductions in design or

Issue Date: 01/13/21 7

2515 Appendix D operational margin, and repetitive work orders could be indicative of licensee acceptance of a long-standing degraded condition.

3.

Any identified deficient condition which may be indicative of equipment tampering.

Inspectors should also evaluate whether licensees actively consider potential for tampering when equipment deficiencies are identified.

4.

Fire hazards that could increase risk, and overall status of fire protection equipment.

5.

Status of on-site and off-site emergency response facilities.

6.

Plant activities which are taking place that may affect the security of the facility such as:

1) security shift turnovers; security officers on posts; 2) security equipment testing and/or review of equipment testing results; 3) security force drills or exercises; and 4) security logs for degraded conditions and compensatory measures. Once a month conduct tours to observe one of these four activities (about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per month should be expended). Guidance for observing these activities is contained in IMC 2201, Appendix D, Facility Status Reviews for Security and Safeguards Inspection Program.

7.

The status of doors to locked high radiation areas and required radiation postings.

Pursue any deficiencies that may impact the Occupational Exposure Control Effectiveness Performance Indicator using IP 71151, Performance Indicator Verification.

8.

Any leakage involving radioactive liquids or gases. Pursue any unmonitored release paths that may impact the Radiological Effluent Occurrence Performance Indicator using IP 71151, Performance Indicator Verification.

9.

Status of remote or alternate shutdown panel areas, including locally required procedures, materials, or communications equipment needed to perform any required actions from these areas.

10. Signs of personnel fatigue or impaired individual alertness which could create a reasonable doubt that an individual is fit to safely and competently perform his or her duties. This applies to all personnel that are granted unescorted access to nuclear power reactor protected areas and individuals that are required to physically report to the licensees Technical Support Center or Emergency Operations Facility by licensee emergency plans and procedures.

07.04 Reactor Safety/Plant Security Interface The events of September 11, 2001 led to significant changes in the security programs at nuclear power plants. With the increased attention to security, we have also recognized that the maintenance of both plant security and safety requires coordination of activities. Such coordination is needed to ensure that actions taken to address security concerns do not adversely affect safety, including emergency preparedness, and that maintenance, operations, or engineering activities do not introduce security concerns. Examples include:

the addition of locks or other barriers to improve security that impedes the ability of operators to take actions included in emergency operating procedures

Issue Date: 01/13/21 8

2515 Appendix D maintenance or construction activity that interferes with security barriers or intrusion detection devices temporary conditions warranting compensatory measures from either security or operations because the conditions differ significantly from plant or risk profiles assumed in either the operating or security procedures changes in site layouts, ingress or egress routes, or security procedures that affect emergency preparedness in areas such as emergency response facility access, emergency preparedness equipment access, site assembly, or staff augmentation times In observing security activities and especially the addition or modification of security features, the inspector should consider and, as appropriate, question the licensee regarding possible safety/security interface issues. In particular, the inspector should look for changes that might adversely affect systems, structures, or operator actions credited in:

Traditional Licensing & Design Bases Functions (e.g., accident analysis, station black out, fire protection programs)

Emergency Operating Procedures Severe Accident Management Guidelines Probabilistic Risk Assessments Radiation Protection Emergency Plan & Emergency Plan Implementing Procedures In observing plant activities such as maintenance, operations, emergency preparedness, and engineering, the inspector should consider and, as appropriate, question the licensee regarding possible safety/security interface issues. In particular, the inspector should look for changes that might adversely affect:

barriers and fences intrusion detection systems alarm and communication systems security event response assumptions for and access to readily available equipment for responding to conditions described in each plants mitigating strategies table modification to equipment relied on in the Emergency Action Level scheme changes to set points contained in the Emergency Action Level scheme 07.05 Construction and Decommissioning Activities Near Operating Units Construction and decommissioning activities may occur in the vicinity of operating unit(s).

Resident inspectors assigned to operating unit(s) should be aware of such activities which may affect the operating unit(s) safety systems. The impact of such activities on operating units will depend on multiple factors such as activity distance from the operating unit(s) and the number of shared SSCs between operating unit(s) and units under construction/decommissioning.

Construction activities for a nuclear power plant may not commence without a construction permit, limited work authorization, or combined license. Activities that are construction are defined by 10 CFR 50.10(a). An applicant may undertake a range of pre-construction activities without a construction permit, limited work authorization, or combined license as defined in 10 CFR 50.10(a)(2). Resident inspectors should contact their regional management regarding any

Issue Date: 01/13/21 9

2515 Appendix D questionable activities to ensure unauthorized construction activities are not occurring and do not have an impact on risk significant systems.

The operating licensee will continue to have the responsibility for ensuring that any site activities do not impact risk significant systems to the operating unit(s) or Independent Spent Fuel Storage Installation (ISFSI). The operating licensees are required to consider the impacts on various aspects of operations, such as emergency planning, radiological protection, security, and demolition of existing facilities and structures. At multi-unit sites, a unit under construction is required, in part, to have managerial and administrative controls to provide assurance that the LCOs are not exceeded as a result of construction activities (10 CFR 52.79(a)(31)). The resident inspectors should understand how construction issues, that may affect the operating unit, are communicated and dispositioned at the operating unit(s).

Resident inspectors should be alert to situations with potential adverse impact (e.g., unit transients or reactor trips). Understanding potential adverse impacts from the pre-construction, construction, and/or decommissioning activities should allow the inspector to implement the appropriate IP to evaluate the potential risk significant systems. Examples of how construction/decommissioning activities could adversely impact the operating unit(s) include, but are not limited to, the following:

1. effects on seismic monitoring from sheet piling installation or explosives used during excavation

2. damage to underground piping, electrical cables, fiber-optics, and telecommunications during excavation or movement of heavy loads

3. disruptions in the switchyard and electrical transmission and distribution systems during movement of heavy loads and associated crane operations

4. emergency preparedness affected by traffic issues from additional personnel for construction/decommissioning or movement of heavy loads

5. heatsink, coolant reservoir intake and discharge structures or piping for the operating unit(s) affected by construction/decommissioning activities occurring near the structures

6. construction/decommissioning activities that cause damage or unexpected changes to the operating unit(s) physical protection systems or components which render those systems nonfunctional

7. fire protection plan impacted by construction/decommissioning activities preventing operator actions through the unavailability of equipment or limited methods to access equipment locations

8. material or debris from the construction/decommissioning site that could impact the operating unit(s) SSCs, switch yard, or off-site power supplies during extreme weather conditions

9. wrong unit maintenance or work activity

10. ISFSI impacted by construction activities.

Issue Date: 01/13/21 10 2515 Appendix D

11. temporary equipment, buildings, debris, lay-down yards that adversely affect yard drainage systems or disrupt flow of floodwater off the site for either operating units or units under construction/decommissioning Multiple heavy load movement and associated crane operation evolutions are expected to occur during pre-construction and construction activities in and around operating unit(s) which have the potential to affect risk significant SSCs. The potential adverse impact of these evolutions on operating unit(s) depends on the distance from vital SSCs and buried piping, the number access roads supporting pre-construction and construction activities and operating unit(s), and if they occur near transmission lines providing off-site power to operating unit(s) and switch yards.

resident inspectors (RIs) should be aware of licensees plans for heavy load movements and crane operations. Additionally, concrete booms can extend to heights where impact with off-site power lines maybe a concern. Additional guidance can be found in Operating Experience Smart Sample FY2007 03, Revision 2, Crane and Heavy Lift Inspection, Supplemental Guidance for IP 71111.20, September 12, 2008, and NUREG 0612, Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A 36, issued July 1980.

Changes to the operating reactors security program may require that the licensee submit revisions to its security plan in accordance with 10 CFR 50.4, Written Communications, 10 CFR 50.54, Conditions of Licenses, or 10 CFR 50.90, Application for Amendment of License, Construction Permit, or Early Site Permit. Examples include the following:

1.

pre-construction, construction, and decommissioning activities that have the potential to affect the integrity, functionality, or performance effectiveness of the operating units security response capability, fields-of-fire, security barriers, illumination capabilities, intrusion detection systems or devices, and access control measures

2.

pre-construction, construction, and decommissioning activities that are conducted in areas where access control is performed for the operating reactor

3.

pre-construction, construction, and decommissioning activities performed in areas adjacent to or in isolation zones of the operating reactor that limit the ability of the operating reactors security force to detect, assess, and interdict potential threats

4.

pre-construction, construction, and decommissioning activities that could prevent operator actions through the unavailability of equipment or limited methods to access equipment locations

5.

temporary conditions warranting compensatory measures from either security or operations because the conditions differ significantly from plant or risk profiles in either the operating or security procedures

6.

changes in site layouts, ingress or egress routes, or security procedures that affect emergency preparedness in areas such as site assembly, staff augmentation times, or accountability of construction personnel.

07.06 Problem Identification Periodically observe licensee management's review of plant deficiencies by attending meetings such as the plant operations review committee and off-site nuclear review board meetings. In addition, most licensees hold Nuclear Safety Culture Monitoring Panels at some periodicity

Issue Date: 01/13/21 11 2515 Appendix D throughout the year (usually quarterly). These meetings allow the licensee to self-assess and determine if there are safety culture or Safety Conscious Work Environment issues before they become larger problems. Inspectors should consider attending Nuclear Safety Culture Monitoring Panels. The inspector should be knowledgeable of major findings from licensee self-assessment activities.

07.07 Resource Estimate The yearly resource expenditures for plant status activities are estimated to be on average: 641 hours0.00742 days <br />0.178 hours <br />0.00106 weeks <br />2.439005e-4 months <br /> for a single-unit site; 699 hours0.00809 days <br />0.194 hours <br />0.00116 weeks <br />2.659695e-4 months <br /> for a dual-unit site; and 908 hours0.0105 days <br />0.252 hours <br />0.0015 weeks <br />3.45494e-4 months <br /> for a triple-unit site.

Resource estimate for Vogtle Unit 1 & Unit 2 is 699 hours0.00809 days <br />0.194 hours <br />0.00116 weeks <br />2.659695e-4 months <br /> total. Resource estimate for Vogtle Unit 3 & Unit 4 is 699 hours0.00809 days <br />0.194 hours <br />0.00116 weeks <br />2.659695e-4 months <br /> total. These yearly resource expenditures include 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per year for resident inspector observations of security-related activities. Time spent conducting security-related activities should be charged to the appropriate quarterly resident inspector inspection report number. Additionally, time expended conducting these activities should be charged to code PS (plant status).

2515D-08 REFERENCES IMC 0327, Steam Generator Tube Primary-to-Secondary Leakage IMC 2201, Appendix D, Facility Status Reviews for Security and Safeguards Inspection Program IMC 2515, Light-Water Reactor Inspection Program - Operations Phase IP 71111.04, Equipment Alignment IP 71111.15, Operability Determinations and Functionality Assessments IP 71111.18, Plant Modifications IP 71111.22, Surveillance Testing IP 71151, Performance Indicator Verification IP 71153, Follow-up of Events and Notices of Enforcement Discretion Regulatory Issue Summary (RIS) 2007-21, Adherence to Licensed Power Limits, Revision 1 (ML090220365)

NUREG-0612, Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A 36, July 1980 (ML070250180)

Operating Experience Smart Sample FY2007 03, Revision 3, Crane and Heavy Lift Inspection, Supplemental Guidance for IP 71111.20 and IP 71111.13, September 1, 2018 (ML18151A450)

END

Issue Date: 01/13/21 Att1-1 2515 Appendix D Assessing Reactor Coolant System (RCS) Unidentified Leakage Rate Trend In order to track and assess the unidentified leak rate trend, the inspector should utilize licensees RCS leakage rate data. Once each month, the inspector should obtain the mean value () and the standard deviation () of RCS unidentified leakage rate for the past three months, representing a 3-month rolling data set, using the Excel spreadsheet (see section titled, Forms, Templates, Sample Reports & More, on Reactor Oversight Process Digital City Web link: https://usnrc.sharepoint.com/sites/NRR-DRO/SitePages/ROP-Digital-City.aspx). During the ensuing month, the inspector should use the resulting and to establish action thresholds as described below.

Note: For licensees who calculate the leak rate more than once per day, ensure that the leak rate value for calculating the mean value is the average for that day. When starting a new operating cycle after refueling, a weekly rolling data set (i.e., most recent 7-day average) of leakage values will be analyzed to determine if the licensee has identified and corrected all potential leakage source(s). Once 3 months of data have been collected, the mean, standard deviation and action levels should be calculated using the Excel spreadsheets listed above.

The mean value () and the standard deviation () are defined by the following equations:

µ = (x1 + x2 +... +xn)/n; = (xi - µ)2/n assuming the unidentified leakage rate, x, is a random variable which has a mean value, µ and a known standard deviation,.

Once a month, the inspector should use the mean value () and the standard deviation

() from the previous three months to calculate the three action level triggers (, µ + 2, µ + 3).

The action levels were determined by statistical analysis:

Action Level I:

Nine (9) consecutive leakage measurements above the µ Action Level II:

Three (3) consecutive measurements exceed the µ + 2 Action Level III:

Two (2) consecutive measurements exceed the µ +3 During the daily plant status review, the inspector should compare the licensee calculated RCS unidentified leakage rate data to the three action level triggers identified below to determine if there is a potential adverse trend and take appropriate actions, if necessary. If the licensee performs the RCS leakage rate calculations several times a day, the inspector should only compare the average positive value per day to the action level triggers. If the licensee, in following its TS, only performs an RCS leakage rate calculation once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then the inspector should perform this comparison once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. For BWRs, if the drywell floor sump is pumped less frequently than daily, then average positive value should only be entered for those days that the sump is actually pumped. Zero or negative values should be entered into the spreadsheet as zero.

Issue Date: 01/13/21 Att1-2 2515 Appendix D Upon exceeding one of the action level triggers, the inspector should consider the licensee in the appropriate action level until the licensee is able to identify, isolate, or repair the leak.

Action Level II - Three (3) consecutive measurements exceed the µ + 2 Actions:

1. Take the steps in Action Level I, if not already done.

2. Review containment data such as sump chemistry samples, pump seal pressures and temperatures (recirculation pumps (BWRs), reactor coolant pumps (PWRs), control rod drive temperature (BWRs), containment atmosphere temperature, pressure, radioactivity, humidity levels, etc.) to determine if source can be attributed to actual RCS leakage.

3. If RCS leakage is confirmed, review licensees plans for identifying source of unidentified leakage and proposed corrective actions.

4. Discuss licensees actions with regional branch chief and engage licensee as necessary.

Action Level I - Nine (9) consecutive leakage measurements above the µ Actions:

1. Assess licensees actions to ensure containment parameters are appropriately being monitoring in accordance with established site-specific procedures.

2. Discuss licensees initial actions with regional branch chief.

Issue Date: 01/13/21 Att1-3 2515 Appendix D END Action Level III - Two (2) consecutive measurements exceed the µ + 3 Actions:

1. Take the steps in Action Level II, if not already done.

2. Discuss increasing trend with licensee management and continue to monitor licensees actions.

3. Ensure regional management at the Director level is informed via the branch chief of the status of licensees actions.

4. If RCS leakage has been confirmed, the appropriate NRR technical branches are notified by the branch chief via the NRR project manager.

5. The resident inspector provides periodic updates on the RCS leak rate and on the status of licensees actions to regional management, and NRR technical branches via the NRR project manager.

Issue Date: 01/13/21 Att2-1 2515 Appendix D Revision History Sheet for IMC 2515 Appendix D Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A 04/03/00 CN 00-003 Reactivated N/A N/A N/A ML020380441 01/17/02 CN 02-001 Revised N/A N/A N/A ML031980014 7/10/03 CN 03-024 Revised to add a statement to remind resident inspectors to periodically check Part 9900 of the inspection manual to keep current on reporting requirements.

N/A N/A N/A ML032660080 9/09/03 CN 03-033 Revised to provide improved guidance to an inspector on the requirement to inform the Materials and Chemical Engineering Branch, NRR, of steam generator tube leaks of greater than three gallons per day.

N/A N/A N/A ML041340196 5/11/04 CN 04-013 Added guidance for reviewing RCS leakage monitoring. Also, requirement to monitor licensee actions when in multiple TS action statements. New requirement to review licensee corrective action summary reports.

N/A N/A N/A ML050410038 1/26/05 CN 05-003 Added more detail to requirement for RCS leakage monitoring.

N/A N/A N/A ML053330037 12/2/05 CN 05-032 Additional clarification to guidance on RCS unidentified leakage trending. Resource estimate for Plant Status has been increased.

N/A N/A

Issue Date: 01/13/21 Att2-2 2515 Appendix D Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A ML061640398 01/26/07 CN 07-004 Included reference to IP 61706 for evaluating reactor power fluctuations (FF 2515D-945). Revised Plant Status resource estimate. Added guidance to inspectors on being sensitive to licensees actions taken to address security concerns do not adversely affect reactor safety and emergency preparedness.

Likewise, licensees actions taken to address reactor safety concerns do not adversely affect plant security (FF 2515-D-998).

N/A ML063460228 N/A ML070740557 04/04/07 CN 07-012 This IMC has been revised to update the RCS unidentified leakage rate spreadsheet web page links. Spreadsheets were updated and converted from Quattro Pro to Excel.

N/A N/A N/A ML080310091 05/01/08 CN 08-014 Revised to include checking for online maintenance activities that could interfere with SSCs and added leakage trending for the first 3 months after the start of a refueling cycle. This revision addresses feedback forms 2515-D-1157 and 2515-D-1178.

N/A N/A N/A ML082110297 09/03/08 CN 08-025 Revised to address lessons learned from severe corrosion of essential service water piping risers at Byron plant (see Operating Experience posting of 10/23/2007) as documented in FF 2515D-1214. Also, incorporated recommendations from FFs 2515D-1156 and 1258 to clarify how to charge for inspection resources used to support facility status reviews for the Security and Safeguards Inspection Program (SSIP) and to make inspectors aware of Plant Status procedure for SSIP (IMC 2201 Appendix D).

N/A ML082410742

Issue Date: 01/13/21 Att2-3 2515 Appendix D Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A ML092180291 11/09/09 CN 09-026 Revised to add guidance for inspectors to look for indications of fatigue when performing plant status reviews. The guidance also provides a reference to new inspection guidance in IP 93002.

Yes 6/17/2009 N/A N/A ML093220843 02/02/10 CN 10-004 Added requirement to have resident inspectors conduct quarterly tours of security-related areas as recommended by CY 2009 ROP realignment process (ML092090312). Increased inspection resources allocated to Plant Status procedure by 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per year to conduct these additional tours of security-related areas by resident inspectors.

N/A ML100070084 N/A ML11279A083 02/24/12 CN 12-003 Provided guidance to be sensitive to deficient equipment conditions which may have resulted from tampering by personnel. Also, made changes to address regional comments associated with feedback forms 1308; 1423; and 1624.

N/A ML12027A113 N/A ML15128A229 09/04/15 CN 15-016 Changes include revisions to (1) power limit reference guidance, (2)

RCS unidentified leakage action levels, and (3) ensure awareness of installation of temporary services.

Feedback forms incorporated into this revision: 2078, and 2141.

Feedback forms reviewed but not incorporated: 2122 and 2131.

N/A ML15187A245 2515D-2078 ML15246A008 2515D-2141 ML15246A009 N/A ML16111B120 04/27/16 CN 16-011 The Action Level triggers described in Attachment 1 were updated to incorporate Regional feedback and the changes recommended in Feedback Form 2141 (and previously adopted through CN 15-016) have been rescinded. [NOTE: The version of this IMC issued on 9/4/15 with an effective date of July 1, 2016 was not implemented and has been superseded by this version].

N/A ML16112A026

Issue Date: 01/13/21 Att2-4 2515 Appendix D Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A ML17152A186 08/25/17 CN 17-016 Revised to indicate that control room walkdown includes review of any compensatory measures in place.

Feedback forms incorporated into this revision: 2230.

Feedback forms reviewed but not incorporated: 2222.

N/A ML17156A218 2515D-2230 ML17178A039 N/A ML17264A782 10/03/17 CN 17-020 Revised to address issues identified IMC 307 peer review (ML16260A079 & ML17047A602). This proposed revision was agreed upon by all members present at the Fall 2017 Reactor Oversight Process Branch Chief Counterpart Meeting. All members also indicated that there is no need for a comment period and the proposed revision can be issued as final.

N/A N/A N/A ML18134A177 11/19/18 CN 18-039 Revised to address: (1) comments provided by NRR/DMLR related to the period of extended operation, (2) Feedback Forms 9900-2273 and 71111.08-2275 which resulted in creation of new IMC 0327, Steam Generator Tube Primary-To-Secondary Leakage, (3) use of mandatory and discretionary language with regards to steam generator tube primary-to-secondary leakage reports to NRR, and (4) a recommendation from the working group established to update the ROP for regulatory actions taken following the Fukushima Dai-ichi accident (ML17164A285).

N/A ML18179A040 N/A ML20238B970 10/05/20 CN 20-046 Revisions are made to reflect resources specifically for Vogtle 3 & 4 as identified in SECY-20-0050, Planned Revisions To The Baseline Inspection Program For The AP1000 Reactor Design, (ML20058F491).

None ML20239A734

Issue Date: 01/13/21 Att2-5 2515 Appendix D Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A ML20323A037 01/13/21 CN 21-004 Consolidated IMC 2515, Appendix F, Reactor Construction Activities Near Operating Unit(s), into this IMC. Incorporated FF 2515D-2412 to add Nuclear Safety Culture Monitoring Panels as a meeting that should be attended. Incorporated guidance regarding Open Phase Condition monitoring.

None ML20325A208 2515D-2412 ML20325A226